Draft 

INTERIM SUMMARY REPORT ON THE EVALUATION OF 

NIOSH'S IDAHO NATIONAL LABORATORY SEC-00219 

PETITION EVALUATION REPORT 

Contract No. 211-2014-58081 

Prepared by 

Bob Barton 

Ron Buchanan 

Joe Fitzgerald 

Milton Gorden 

John Mauro 

Amy Meldrum 

Stephen Ostrow 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

July 2015 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted.
Effective Date: 

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S. Cohen & Associates: Document No.: 

Interim Summary Report on 

Technical Support for the Advisory Board on INL SEC-00219 ER 

Radiation & Worker Health Review of Effective Date: 

NIOSH Dose Reconstruction Program Draft – July 6, 2015 

Revision No.: 

0 (Draft) 

Interim Summary Report on the Evaluation of 

NIOSH's Idaho National Laboratory SEC-00219 Page 2 of 34

Petition Evaluation Report 

Task Manager: Supersedes: 

_____________________________ Date: ___________ N/A 

John Stiver, MS, CHP 

Project Manager: Reviewers: 

 Joe Fitzgerald _____________________________ Date: ___________ 

 Stephen Ostrow John Stiver, MS, CHP 

 John Stiver 

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

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TABLE OF CONTENTS 

Abbreviations and Acronyms ..........................................................................................................5

1.0 Introduction and Background ..............................................................................................7

2.0 SC&A Assessments ...........................................................................................................10

2.1 Class Definition .....................................................................................................10

2.1.1 Interview Summaries .................................................................................10

2.2 Dose Reconstructability: Data Completeness, Data Quality, and Data 

Adequacy ...............................................................................................................11

2.2.1 Fission and Activation Product (FAP) Bioassay Indicator 

Radionuclides .............................................................................................11

2.2.2 Reactor Modeling.......................................................................................14

2.2.3 Burial Grounds ...........................................................................................16

2.2.4 Central Facilities Area (CFA) ....................................................................20

2.2.5 Chemical Processing Plant (CPP) Pre-1963 ..............................................22

2.2.6 Test Area North (TAN) ..............................................................................25

3.0 References ..........................................................................................................................33
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LIST OF TABLES 

Table 1. INL SEC 00219 Evaluation Report: Feasibility of Reconstructing Internal Doses 

for 1/1/63–12/31/74 ......................................................................................................... 9

Table 2. ORIGEN-S Irradiation Parameters for the Nine Representative Reactor Cases ........... 14

Table 3. INL Test Reactor Area (TRA) Reactors and Support Facilities .................................... 15

Table 4. CFA Facilities of Concern ............................................................................................. 20

Table 5. Feasibility Summary for CPP (1953–1974)................................................................... 23

Table 6. Documents Found for TAN in SRDB with Detailed Dosimetry Records ..................... 27

Table 7. Visitor or Summary Documents Found for TAN in SRDB ........................................... 31

LIST OF FIGURES 

Figure 1. Number of Contaminated Surfaces in CPP-601 Sample/Operating Corridors and 

Access Corridor ............................................................................................................. 24

Figure 2. Dates of Dosimeters Available in SRDB Documents for All of TAN ......................... 29

Figure 3. Dates of Dosimeters for 1956 taken from Figure 2 ...................................................... 29

Figure 4. Dates of Dosimeters for Various Sub-Areas of TAN ................................................... 30

Figure 5. Number and Type of Documents Found in the SRDB According to the Year(s) 

They Cover .................................................................................................................... 31
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ABBREVIATIONS AND ACRONYMS 

Advisory Board 

or ABRWH Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

ANC Aeroject Nuclear Company 

ANL-W Argonne National Laboratory West 

ANP Advanced Nuclear Propulsion 

ARMF Advanced Reactivity Measurement Facility 

ATR Advanced Test Reactor

ATRCF Advanced Test Reactor Critical Facility 

CAM constant air monitor 

CATI Computer-Assisted Telephone Interview 

CFA Central Facilities Area 

c/m counts per minute 

CPP Chemical Processing Plant 

CFRMF Coupled Fast Reactivity Measurement Facility 

D&D decontamination and decommissioning 

DCAS Division of Compensation Analysis and Support 

DOE (U.S.) Department of Energy 

DR dose reconstruction 

DU depleted uranium

ER Evaluation Report 

ERDA Energy Research Development Administration 

ETR Engineering Test Reactor 

ETRCF Engineering Test Reactor Critical Facility 

FAP Fission and Activation Product

FFTF Fast Flux Test Reactor 

IET Initial Engine Test

IMBA Integrated Modules for Bioassay Analysis 

INC Idaho Nuclear Corporation 

INL Idaho National Laboratory 

HP Health Physics 

LPTF Low Power Test Facility 

MAP mixed activation product 

MDA minimum detectable activity 

MFP mixed fission product
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mrem millirem 

MTR Materials Test Reactor 

MTHM metric ton of heavy metal 

MTU metric ton unit 

MW megawatt 

MWd megawatt-day 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH/OCAS Claims Tracking System 

NRC Nuclear Regulatory Commission 

NRTS National Reactor Testing Station 

OCAS Office of Compensation Analysis and Support (now DCAS) 

ORAUT Oak Ridge Associated Universities Team 

ORIGEN2 Oak Ridge Isotope Generation and Depletion Code 

ORNL Oak Ridge National Laboratory 

R Roentgen 

RFP Rocky Flats Plant

RMF Reactivity Measurement Facility 

RML Radiation Measurement Laboratory 

RWMC Radioactive Waste Management Complex 

SC&A S. Cohen and Associates 

SEC Special Exposure Cohort 

SMC Specific Manufacturing Capability 

SRDB Site Research Database 

STPF Shield Test Pool Facility 

TAN Test Area North 

TBD Technical Basis Document 

TIB Technical Information Bulletin

TLD thermoluminescent dosimeter 

TRA Test Reactor Area 

TRU transuranic 

TSF Technical Support Facility 

WBC whole-body count 

WG Work Group 

WRRTF Water Reactor Research Test Facility 
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1.0 INTRODUCTION AND BACKGROUND 

Following the National Institute for Occupational Safety and Health’s (NIOSH’s) release of the 

Idaho National Laboratory (INL) Special Exposure Cohort (SEC) Petition SEC-00219

Evaluation Report (ER) on March 12, 2015 (NIOSH 2015), and its presentation at the March 26, 

2015, Advisory Board on Radiation and Worker Health (ABRWH, or the Advisory Board) 

meeting in Richland, Washington, SC&A was asked to begin reviewing the ER. The April 22, 

2015, planning teleconference held with SC&A, NIOSH (including its contractor), and some 

INL Work Group (WG) members, as well as several e-mail communications, established the 

scope of SC&A’s preliminary investigations, which are intended initially to support a July 8, 

2015, INL WG meeting and a July 23, 2015, session at the ABRWH meeting in Idaho Falls, 

Idaho. The results of the discussions at these meetings will lead to further investigations to 

support the Board’s consideration of the SEC petition and NIOSH’s evaluation report. 

In light of the complex nature of the site, its operations, and the SEC ER itself, the Advisory 

Board determined that its review of the ER should be performed in a graded manner, whereby 

SC&A would first conduct a preliminary review of certain issues of immediate concern to the 

Board. It is recognized that this review, in addition to being preliminary due to time constraints, 

also is very much a work in progress, as NIOSH has acknowledged that the evaluation itself is 

still a work in progress as new information continues to be obtained, more worker interviews 

conducted, and coworker and other dose reconstruction (DR) models are developed. 

SC&A’s initial charge by the Board was to provide information and assessments, primarily on 

two topics of concern to the WG: 

(1) Class Definition: NIOSH proposed the following class definition in its ER: 

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Idaho National Laboratory in 

Scoville, Idaho, and were monitored for external radiation at the Idaho Chemical 

Processing Plant (CPP) (e.g., at least one film badge or TLD dosimeter from 

CPP) between January 1, 1963 and December 31, 1974 for a number of work 

days aggregating at least 250 work days, occurring either solely under this 

employment, or in combination with work days within the parameters established 

for one or more other classes of employees in the Special Exposure Cohort. 

The above definition implicitly assumes that all employees who might have been present at CPP 

during the class period and potentially exposed to internal sources of radiation (which exposure 

NIOSH claims it cannot reconstruct for CPP during the class period) can be identified by the fact 

that they were all monitored for external radiation, and that their exposures can be attributed to 

CPP. As stated in the ER (p. 6): “By policy, INL monitored by dosimeter all personnel who 

were expected to receive any radiation dose or whose work was centered at the site.” The 

specific concerns for this ER are whether all workers exposed to radiation at CPP were, in fact, 

monitored; whether their dosimetry records are available; and whether the workers affected by 

this SEC can be reliably placed at the CPP during the specified time period. 
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SC&A considered the class definition assumption through two lines of investigation: by 

reviewing about 60 workers’ statements made through three series of interviews conducted 

during June, September, and November 2014; and by performing a detailed examination of the 

available dosimetry and location records supporting identification of which workers were 

exposed at CPP during the given time period (this examination may be referred to as a “gap 

analysis,” since it is attempting to identify any “gaps” in the available data). That analysis 

(SC&A 2015) was delivered to the INL WG as a stand-alone document on June 29, 2015. 

(2) Dose Reconstructability and Gap Analysis: The Board recognized that NIOSH’s 

investigations regarding reserved areas (see Table 1) are under way, and it would be 

premature for the Board and its contractor to perform SEC assessments into such areas 

before NIOSH had an opportunity to complete those investigations, or at least portions of 

those investigations. Hence, this report does not address issues that are currently held in 

reserve by NIOSH. However, the Board did conclude that it would be appropriate for 

SC&A to begin its investigations into activities, facilities, and time periods at the site that 

NIOSH recommended denial of an SEC, because it felt that worker doses at such 

facilities could be reconstructed with sufficient accuracy. The Board also recognized that 

the scope of such an investigation was vast and should be approached in a highly 

organized and deliberative manner. As such, the Board directed SC&A to prepare a plan 

for identifying data gaps for each of the areas, facilities, and campaigns at the site where 

NIOSH felt that doses could be reconstructed with sufficient accuracy. This status update 

report is provided in response to that Board directive. The report represents and plans for 

a data gap analysis, and also begins implementing the plan for specific sites as a means to 

help determine that such a plan will be useful in assisting the Board in its SEC 

recommendations.

While the SEC class definition includes a specific cohort of workers over a specific time period, 

at the same time it also excludes other INL workers (i.e., those working at locations other than at 

CPP) at other time periods. NIOSH claims that it can adequately reconstruct doses for those 

workers and, so, did not include them in the SEC class. SC&A approached this issue through 

several different methods that can be characterized as either horizontal or vertical. In the former, 

SC&A examined the DR methodology applied by NIOSH for all INL personnel, and for the 

latter SC&A looked at some specific characteristics of the individual areas at the INL site where 

personnel might have been exposed to radiation. The working areas considered are: 

 Chemical Processing Plant (CPP) 

 Test Reactor Area (TRA) 

 Test Area North (TAN) 

 Misc. Reactor Areas 

 Central Facilities Area (CFA) 

 Burial Grounds 

Information on the characteristics of these areas appears in many sources, including, in addition 

to the ER, the INL site profile [comprised of six separate technical basis documents (TBDs)], 

reports produced by NIOSH (and its contractor) and SC&A, documents collected in the Site 
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Reference Data Base (SRDB), as well as documents appearing elsewhere. SC&A has examined 

some pertinent documents in this preliminary assessment. 

Table 1 summarizes the ER’s conclusions about which combinations of operating areas and time 

periods can be reconstructed and which cannot. Note that certain combinations have been 

“reserved” for further study, which might result in a determination that doses can be 

reconstructed, or a recommendation that affected personnel be added to the SEC class. In 

addition, the feasibility of DR for certain area and time period combinations has been assumed 

based on the assumption of future development of the bioassay database and acceptable 

coworker models. 

For administrative reasons, Argonne National Laboratory West (ANL-W), although physically 

located within the INL site, will be the subject of a separate SEC investigation, which NIOSH is 

currently conducting. Hence, ANL-W has been excluded from these discussions. 

Table 1. INL SEC 00219 Evaluation Report: Feasibility of Reconstructing 

Internal Doses for 1/1/63–12/31/74 (a),(b)

Operating Area (c) Reconstruct?

Chemical Processing 

Plant (CPP)

Yes: 1/1/53–12/31/62: Bioassay data available. 

No: 1/1/63–12/31/74: Insufficient bioassay data to support reconstruction of internal 

exposures to U, Np, Pu, and other related transuranics. Beginning in 1963, increased 

α-contamination levels were detected with no accompanying increase in Pu bioassay.

Reserved: 1/1/75 – ?: Evaluate CPP HP Program Improvements 1971–1980.

Test Reactor Area (TRA) Yes:

 1/1/52–12/31/66

 1/1/67–12/31/70: coworker model.

(d)

Test Area North (TAN) Yes: 

 1/1/55–12/31/66

 1/1/67–12/31/70: coworker model.

Reserved: 1/1/61–12/31/70: TAN-607 (Fuel Storage Vaults) and TAN-615 (Actuator 

Building) due to potential U exposures in the absence of MFPs.

Misc Reactor Areas Yes:

 1/1/55–12/31/66

 1/1/67–12/31/67 and 1/1/69–12/31/70: coworker model.

Reserved: 1/1/68–12/31/68: Insufficient data currently available for ARA-I (Auxiliary 

Reactor Area-I) due to potential unmonitored exposures to Pa-233 from separation of 

that isotope from irradiated Th slugs. 

Central Facilities Area 

(CFA)

Yes: 

 1/1/49–12/31/66

 1/1/67–12/31/70: coworker model.

Burial Grounds Yes:

 1/1/52–12/31/66

 1/1/67–12/31/68: coworker model.

Reserved: 1/1/69–12/31/70: Evaluate a newly-implemented procedure of waste 

exhumation and retrieval.

(a) SEC 00219 time period evaluated by NIOSH is 1/1/63–12/31/74 for CPP based on insufficient data to 

reconstruct internal doses. 

(b) Operations at INL began 1/1/49, but some of the areas began operations later. 

(c) ANL-W is not included in this SEC, but will be the subject of SEC 00224, which is under development. 

(d) NIOSH is developing an internal, mixed fission product (MFP) coworker model stratified by area for time 

periods beginning 1/1/67.
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2.0 SC&A ASSESSMENTS 

SC&A addressed the issue of class definition as requested by the Board in a recently issued 

paper (SC&A 2015), and has addressed our traditional review of the INL ER by an early focus 

on significant issues and gaps for those operations and time periods for which dose 

reconstructability is being claimed. This latter review is being performed through a number of 

relatively quick studies, which are a combination of scoping analyses and selected “vertical” 

investigations. Since the studies are still in progress, and recognizing that the Board would like 

to preview some of the results, though preliminary, for discussion at the July 8, 2015, INL WG 

meeting and the July 23, 2015, INL SEC session at the ABRWH meeting, SC&A has combined 

summaries of these individual draft studies into this interim summary report. Some of the 

reports and studies are concerned with specific INL areas, while others are concerned with 

common methodologies and concerns that crosscut the different operating areas. Upon 

completion of the individual studies, SC&A will provide a final report that combines all aspects 

of these studies, and our overall preliminary findings, or a series of individual reports. The 

progress to date of all investigations is summarized in the following sections. 

2.1 CLASS DEFINITION 

SC&A addressed the issue of class definition with two very different approaches; the first 

combed through worker interview summaries and the second examined in depth the availability 

of worker records pertaining to external dosimetry at the CPP. As previously noted, the detailed 

review of external dosimetry was reported in the stand-alone document SC&A 2015. The 

description of the interview summaries is discussed in this status update report. 

2.1.1 Interview Summaries 

SC&A reviewed about 50 sets of worker interview summaries that were conducted by SC&A, 

NIOSH, and Board members in Idaho in June, September, and November 2014. At the time of 

the review, not all of the summaries examined were finalized. SC&A believes that none of the 

summaries contradict NIOSH’s SEC class assumption that all workers at CPP had personal 

monitors for external radiation during the proposed SEC time period of January 1, 1963–

December 31, 1974. Of course, it should be realized that absence of a contradictory statement in 

the interview summaries does not confirm the validity of the SEC class definition, since not all 

workers at the plant were interviewed and memories can be faulty. 

Several statements indicate that the requirement for all workers at CPP to be badged upon entry 

to the area might have been relaxed at some later time (i.e., later than the end date of the 

proposed SEC time period) to require badging only for entry into radioactive areas, or area 

monitoring substituted for personal monitoring in some cases (which might have been only for 

construction in “new” areas outside the radioactive boundaries). This could be an issue if 

NIOSH decides to extend the time period of the SEC class beyond the current December 31, 

1974 endpoint. 

Recommendation: SC&A recommends that this line of inquiry be continued to include both past 

interviews that were not available when SC&A did its initial review, as well as new interviews 
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where more specific questions could be asked of the workers regarding external dosimetry for 

CPP workers. 

2.2 DOSE RECONSTRUCTABILITY: DATA COMPLETENESS, DATA QUALITY, 

AND DATA ADEQUACY 

The second major issue examined in this preliminary assessment of the INL SEC-00219 ER

consists of initial investigations into whether NIOSH can adequately reconstruct doses as it 

claims for the areas and time periods outside the proposed SEC class definition. These 

investigations are best described as data gap analyses, where SC&A reviewed selected Site 

Research Database (SRDB) folders as a means to identify areas, sub-areas, and time periods at 

INL, including various research projects and special campaigns, where worker external and/or 

internal dosimetry data might be incomplete, or might have employed protocols that could have 

missed potentially important sources of exposures; examples of the latter are internal exposures 

to exotic radionuclides or neutrons. SC&A performed several preliminary studies to examine 

these issues. These investigations are still underway, but are summarized in the following 

sections. 

2.2.1 Fission and Activation Product (FAP) Bioassay Indicator Radionuclides

SC&A assessed the appropriateness of the methods that NIOSH proposes to use to determine 

internal doses for areas and times where it asserts that it can reconstruct doses. NIOSH’s basis 

for assigning internal doses for most years and locations not covered by the proposed SEC class 

and the reserved areas and dates relies on the following four assumptions: 

A. FAP Bioassays – Sufficient workers’ records containing fission and activation product 

(FAP) bioassay (in-vitro and in-vivo) results are available to assign intakes and resulting 

doses from FAP (some periods/areas may need an FAP coworker model developed). 

B. FAP Intakes – Except for special situations, all the dosimetric significant FAP intakes are 

directly tied to an indicator radionuclide (Sr-90 or Cs-137). The FAP ratios and intake 

assignment methods provided in ORAUT-OTIB-0054 [ORAUT 2015] bound all FAP 

exposure potentials at INL. 

C. Actinide Intakes – Except for special situations, the actinides (uranium, plutonium, 

thorium, etc.) intakes are directly tied (in a constant ratio) to the FAP; therefore, actinide 

intakes and resulting doses can be assigned using Table 5-22 (Sr-90 ratios) and/or 

Table 5-23 (Cs-137 ratios) of ORAUT-TKBS-0007-5 [ORAUT 2010]. 

D. Special Situations Actinides – Personnel involved in operations and situations (planned 

or unplanned) with actinides present, that were not directly tied to an FAP in a constant 

ratio, were adequately monitored, and the results are available in the workers’ records. 

Therefore, actinide intakes and resulting doses can be reconstructed in these special 

situations. 

SC&A has evaluated Item B [application of FAP indicator ratios in Tables 7-3 of ORAUTOTIB-0054 (ORAUT 2015) to INL FAP intakes], and Item C [application of actinide indicator 

ratios in Tables 5-22 and 5-23 of ORAUT-TKBS-0007-5 (ORAUT 2010) to INL actinide 
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intakes]. SC&A performed benchmark assessments for some of the ratio values by comparing 

them to actual measured results. SC&A attempted to assess this situation by searching the INL 

claim files in the NIOSH/OCAS Claims Tracking System (NOCTS) and locating claims that 

have FAP and/or actinide exposures with results greater than the minimum detectable activity 

(MDA), and indicator radionuclide data for workers potentially exposed to intakes from reactor 

fuels (not intakes of FAP or actinides that have been separated). 

The following two sections outline the background for the investigations into Items B and C. 

Item B – FAP Intakes Using Indicator Radionuclide Ratios 

Both the ER and the INL internal dosimetry TBD (ORAUT 2010) rely on the methodology of 

ORAUT-OTIB-0054 (ORAUT 2015) to assign FAP intakes using Sr-90 or Cs-137 as the 

indicating radionuclide for numerous other FAP radionuclides for most locations and time 

periods. Specifically, often workers’ bioassays, air samples, or smears were only analyzed for 

Sr-90 by gross beta counting, Cs-127 by gross gamma counting, or Cs-137 by whole-body 

counts (WBCs); other FAPs were not identified or quantified during the process. The general 

use of ORAUT-OTIB-0054 (ORAUT 2015) for assigning FAP intakes from reactor fuels has 

been previously addressed and will not be further evaluated here.1 However, because a wide 

variety of different reactor fuels were used at INL over the years, it is prudent to determine if 

there are benchmark examples where the actual recorded data at INL agree or disagree with the 

values recommended in ORAUT-OTIB-0054 (ORAUT 2015). 

Item C – Actinide Intakes Using Indicator Radionuclide Ratios 

The ER notes that NIOSH assigns actinide intakes (except for certain special situations) directly 

tied in a constant ratio to the FAP through Table 5-22 (Sr-90 ratios) and/or Table 5-23 (Cs-137 

ratios) of the internal dosimetry TBD (ORAUT 2010). Because this method involves assigning 

intakes and doses over a long time span and numerous areas/operations, it results in the majority 

of the alpha internal dose assignments for INL workers. Therefore, it is necessary to evaluate the 

origin of these data in some depth to determine their applicability to the DR process. 

SC&A evaluated the ratio methodology used by NIOSH to determine internal exposures in many 

cases by employing three different strategies. 

1. Claims: SC&A searched the INL claim files on NOCTS to locate claims involving FAP 

and/or actinide exposures, with results greater than the MDA, and indicator radionuclide 

data for workers potentially exposed to intakes from reactor fuels (not intakes of FAP or 

actinides that have been separated). However, search efforts to date of over 6,000 INL 

claims have not located many claims with these criteria. Further investigation along 

these lines is needed. 

 

1

 SC&A extensively evaluated ORAUT-OTIB-0054 (ORAUT 2015) for the Board’s Subcommittee on 

Procedures Review. Of the many documents produced relevant to the TIB, a good, technical presentation of how 

NIOSH modeled several different representative reactors and fuel combinations appears in ORAUT-RPRT-0067 

(ORAUT 2014). 
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2. MDA: SC&A analyzed several scenarios that involved applying the MDA values for 

Sr-90 or Cs-137 and plutonium (the largest actinide alpha dose contributor) to several 

claims and some hypothetical cases. This analysis was performed because it might 

indicate any obvious inconsistencies between plutonium intakes assigned by the 

Integrated Modules for Bioassay Analysis (IMBA) program using the MDA method, 

compared to those using the Pu/Sr or Pu/Cs ratios of Tables 5-22 and 5-23 of ORAUT 

2010 respectively. Preliminary results using this approach were mixed and inconclusive 

to date, but should be pursued further. 

3. SRDB: SC&A performed some keyword searches on the SRDB for documents 

containing positive bioassay, air filter, or smear data. SC&A succeeded in locating four 

relevant documents consisting of a total of approximately 35,000 pages total. The 

recorded data were analyzed to determine FAP and actinide radionuclide ratios compared 

to the indicator radionuclide. 

These investigations resulted in a number of observations and areas of concern, some of which 

require further investigation, including the following. This preliminary investigation indicates 

that for the small sampling of data points in the 1983–1985 time period: 

 FAP intakes derived using OTIB-0054 (ORAUT 2015) would be greater than those that 

would be derived from measured data. 

 Pu-238 intakes derived using ORAUT-TKBS-0007-5 (ORAUT 2010), Tables 5-22 or 

5-23, would be greater than those that would be derived from measured data for nasal 

swabs and a smear. 

 Pu-238 intakes derived using ORAUT-TKBS-0007-5 (ORAUT 2010), Tables 5-22 or 

5-23 would be less than those that would be derived from measured data for CAM filters. 

 Am-241 intakes derived using ORAUT-TKBS-0007-5 (ORAUT 2010), Tables 5-22 or 

5-23 would be less than those that would be derived from measured data for nasal swabs. 

Recommendations: Given these initial observations and areas of concern, we suggest the 

following path forward concerning Item B (FAP) and Item C (actinides): 

 Determine if the burnup in the fuel elements used by NIOSH is applicable/ 

bounding to the situations at INL. 

 Investigate the use of one model and only three fuel elements to bound the 

intakes/doses. 

 Determine if records of analyses of dissolver contents (chopped/shredded fuel 

elements) are available; preferably, for a variety of INL reactor fuel elements. 

 Evaluate NIOSH’s recommended ratio value, especially for actinides. Further 

investigations along these lines may be aided by the electronic bioassay database; 

even if this database is presently incomplete, paired FAP and actinide bioassays 

could provide relevant information. 
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2.2.2 Reactor Modeling 

This section examines the validity of the isotopic ratios of ORAUT-OTIB-0054 (ORAUT 2015),

regarding whether the parameters of the representative reactor cases of the TIB envelope the 

operational parameters of the INL reactors, starting with those in the Test Reactor Area (TRA). 

Future investigations will need to expand to other reactors on the site, such as in Test Area North 

(TAN). The ER cites OTIB-0054 (ORAUT 2015) multiple times. For example: 

At TRA, urine samples were typically analyzed only for gross beta, gross gamma, 

and/or strontium radioactivity. For such samples, NIOSH will assess missed 

Sr-90 and/or Cs-137 intakes in accordance with ORAUT-OTIB-0054 [ORAUT 

2015] and ORAUT-OTIB-0060 [ORAUT 2007]. 

NIOSH uses OTIB-0054 (ORAUT 2015) to determine internal doses using indicator 

radionuclides in cases where only gross beta or gross gamma measurements are available. The 

general validity and applicability of the TIB has been evaluated by SC&A as part of the 

Procedures Review Subcommittee review process and all findings were closed. In brief, after 

initially considering seven reactor cases and after a complicated process involving sets of 

ORIGEN runs (Croff 1980), the downselect process reduced the cases to four representative 

reactors: 

 Advanced Test Reactor (ATR) – High flux reactors 

 Fast Flux Test Facility (FFTF) – Na-cooled fast reactors 

 Hanford N-Reactor – Pu production reactors 

 TRIGA with stainless steel cladding – Research reactors 

Multiple ORIGEN-S runs [ORNL 2015 – ORIGEN-S is part of the SCALE code system 

developed and maintained by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear 

Regulatory Commission (NRC)] produced a total of nine representative cases. OTIB-0054 

(ORAUT 2015), Table 5-2 (reproduced here as Table 2) lists the parameters and bases selected 

for each of the cases. 

Table 2. ORIGEN-S Irradiation Parameters for the 

Nine Representative Reactor Cases

Case Parameters Basis

ATR 1

Specific power = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximum burnup at nominal power.

ATR 2

Specific power = 8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup =314,904 MWd/MTU

Maximum burnup at maximum assembly power.

ATR 3

Specific power = 2,379.1 MW/MTU

Irradiation time = 56 days. 

Burnup = 133,230 MWd/MTU

Nominal burnup at nominal power.

FFTF 1

Specific power = 163.8 MW/MTHM

Irradiation time = 929.4 days

Burnup = 152,230 MWd/MTHM

Maximum burnup at nominal power.
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Table 2. ORIGEN-S Irradiation Parameters for the 

Nine Representative Reactor Cases 

Case Parameters Basis

FFTF 2

Specific power = 163.8 MW/MTHM

Irradiation time = 488.3 days

Burnup = 79,984 MWd/MTHM

Nominal burnup at nominal power.

N Reactor 1

Specific power = 10.4 MW/MTU

Irradiation time = 114.2 days

Burnup = 1,188 MWd/MTU

Production of weapons-grade plutonium (nominal

6% Pu-240 content) at nominal power.

N Reactor 2

Specific power = 10.4 MW/MTU

Irradiation time = 285.6 days

Burnup = 2,970 MWd/MTU

Production of fuel-grade plutonium (nominal 12%

Pu-240 content) at nominal power.

TRIGA 1

Specific power = 15.57 MWd/MTU

Irradiation time = 730.1 days

Burnup = 11,368 MWd/MTU

Maximum burnup at nominal power.

TRIGA 2

Specific power = 15.57 MW/MTU

Irradiation time = 115.2 days

Burnup = 1994 MWd/MTU

Nominal burnup at nominal power.

Source: ORAUT 2015, Table 5-2 

The TIB indicates that NIOSH will customarily consider all nine reactor cases when doing a DR

and, in the absence of individual worker data, might apply data from all four decay times 

(10 days, 40 days, 180 days, and 1 year), as well, if required by a specific DR case.

Given the parameters of the representative reactor cases shown in Table 2, SC&A is 

investigating, starting with TRA, whether all the normal and off-normal operating scenarios of 

the INL reactors are adequately enveloped, so that the ratios given in OTIB-0054 (ORAUT 

2015) will be valid. Table 3 lists the TRA reactors, of which the first three, intended to test 

materials in high neutron flux environments, are by far the largest and most significant. 

Table 3. INL Test Reactor Area (TRA) Reactors and Support Facilities 

Facility Name Building No. First 

Used Last Used Facility Type

Materials Test Reactor (MTR) TRA-603 1952 1970 Reactor

Engineering Test Reactor (ETR) TRA-642 1957 1981 Reactor

Advanced Test Reactor (ATR) TRA-670 1967 In use Reactor

Reactivity Measurement Facility (RMF) TRA-603 1954 1962 Reactor

Advanced Reactivity Measurement Facility No. I

(ARMF-I)

TRA-660 1960 1974 Reactor

Advanced Reactivity Measurement Facility No. 2

(ARMF-II); renamed Coupled Fast Reactivity 

Measurement Facility (CFRMF)

TRA-660 1962 1991 Reactor

Engineering Test Reactor Critical Facility (ETRCF) TRA-654 1957 1982 Reactor

Advanced Test Reactor Critical Facility (ATRCF) TRA-670 1964 In use Reactor

Hot Cell Facility TRA-632 1954 In use Research

Gamma Facility TRA-641 1955 ND Research

Radiation Measurement Laboratory (RML) TRA-604 1952 In use Sampling/ Research

Radiochemistry Laboratories TRA-604 1950 In use Sampling/ Research

Alpha Laboratories TRA-652 1961 1970 Sampling/ Research

HB-4 Crystal Spectrometer TRA-603 1952 1970 Research

Fast (Neutron) Chopper TRA-665 1952 1970 Research
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SC&A’s preliminary examination of the operating scenarios of the three main, high-power 

reactors and their fuel, burnup, and power level combinations indicates that the use of the OTIB0054 (ORAUT 2015) methodology and values appears appropriate. 

Recommendation: SC&A believes that further study should be done for off-normal operations, 

including “special” materials irradiation runs, and for any incidents that might have occurred and 

exposed personnel to radiation. In addition, this line of investigation should be extended to other 

reactor areas, especially TAN, which hosted very unusual reactor experiments, such as the 

Advanced Nuclear Propulsion (ANP) reactors, which were decidedly different in fuel 

composition and arrangement and operation than other types of reactors. 

2.2.3 Burial Grounds 

While mixed fission products (MFPs) and Rocky Flats Plant (RFP) plutonium undoubtedly did 

dominate the radioactive waste being managed at the Burial Grounds, the actual radionuclide 

content of specific onsite and offsite solid waste being handled at any given time was not 

normally known. For example, the onsite NRTS waste was nominally described as “mixed 

fission products,” but could consist of a variety of radioactive constituents (SRDB Ref ID: 

#138957). Likewise, for RFP wastes, shipments did not include paperwork describing physical 

form and radionuclide content until 1964 (RWMC 1985). For nationwide commercial and 

military radioactive waste received at the Burial Grounds in 1960–1963, it was found that these 

offsite shipments arriving at the Burial Grounds did not have documentation that identified the 

waste container contents, and proposed shipments “seldom, if ever, [were] the same as the 

shipment itself” (SRDB Ref ID: #138704). 

Health Physics (HP) personnel were assigned to the Burial Grounds to perform monitoring, but 

also fulfilled a prime operational role for managing the radioactive waste disposal at the Burial 

Grounds. This operational role encompassed initiating and receiving work requests for Burial 

Grounds operation, reporting on costs for operations to the Atomic Energy Commission (AEC), 

developing and recording plot plans for pits and trenches, witnessing and maintaining a log of all 

burials, in addition to routine radiological surveillance functions (Burial Grounds 1962). This 

“dual” responsibility of Idaho Nuclear Corporation’s (INC’s) Health and Safety Branch for both 

operations and safety of the Burial Grounds was found to be an organizational conflict of interest 

and led to a lowered performance rating for INC in an audit conducted by the AEC in 1971.2

Safe work permits were prepared for dumping operations, but a cursory review of those for RFP 

waste disposal from the 1960s shows a wide range of work controls prescribed—from no 

controls or precautions identified (except for steel-toed guards) to a wide range of provisions 

calling for tool checks, hand and foot counting, intermittent surveying, and final monitoring with 

an alpha survey meter (SWP 1962–1964; SRDB Ref ID: #141608). Some of the interview 

accounts by Burial Grounds operators (backed up by contemporary photographs) suggest little in 

 

2

 Due to leaking barrels received at the Burial Grounds from RFP in 1970, the AEC requested that Burial 

Grounds operations be suspended pending an independent internal review of HP procedures, which found this 

organizational conflict to be a contributing problem (INC 1971). 
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the way of protective clothing except for a hardhat and coveralls. The requirements cited in the 

permits seem to be more related to the HP writing them than to the job, itself. 

In terms of contamination control during operations, former Burial Grounds workers noted that 

when RFP TRU waste drums were routinely dumped from trucks into waste pits, the drum lids 

would sometimes pop off, leading to scattering of contents throughout the pit (SRDB Ref ID: 

#141861). An NRTS-wide road contamination survey in 1969 found the following: 

The number and intensity of the contaminated areas along the route to the Burial 

Ground indicates inadequate handling procedures and is a matter of concern. It 

appears that either waste shipped to the Burial Ground has been improperly 

packaged or waste unloaded at the Burial Ground was carelessly handled 

causing contamination of the returning vehicle and then the roadway. Surveys of 

vehicles and equipment leaving the burial area should indicate if the latter is 

occurring. (SRDB Ref ID: #125523) 

This 1969 survey found contamination by various fission product nuclides, including Cs-137 and 

Co-60, with measured dose rates (at contact) as high as 20 and 30 R/hour. 

By their own admission, the capability of the contractor to even administer a valid contamination 

control program is questionable for the Burial Grounds. In an April 5, 1972, internal 

memorandum within Aeroject Nuclear Company (ANC) (ANC 1972), the HP organization 

identified poor and outdated instrumentation due to lack of funding as undercutting its program. 

It is noted that a longstanding problem existed with outmoded HP smear counters at CFA, with 

no smear counting equipment at the Burial Grounds before February 1972. This was considered 

to be “a completely intolerable situation which must be corrected as soon as possible if we [ANC 

HP] are to avoid a serious contamination incident.”3 The HP organization further stressed that 

smear counting was the only capability available to them to detect “loose” contamination and to 

differentiate between beta-gamma and alpha contamination. In its final weighing of potential 

consequences related to the lack of this equipment, ANC HP makes the following assessment: 

The Burial Ground has been operating for years without any smear counting 

equipment and has avoided serious contamination incidents by luck and/or by the 

experience of well-trained health physicists. The Burial Ground operation cannot 

comply with present radioactive on-site shipping regulations because of a lack of 

proper detection equipment. With increased emphasis on contamination control, 

it is absolutely imperative that state of the art simultaneous smear counting 

equipment be purchased for the Burial Ground as a minimum. (ANC 1972) 

In conclusion, the ANC HP manager concludes that, “without smear counting equipment, or 

outdated and inadequate smear counting equipment, there is, at best, only a token effort of 

contamination control.” 

 

3

 It was noted that as an interim measure, at that time in 1972, the ANC HP office would be deploying an 

Eberline RM-14 with HP 210 probe to count smears at the Burial Grounds, and would seek a surplus beta-gamma 

counter as a stop-gap measure. 
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A number of concerns relevant to the SEC petition emerge from the material quoted above, as 

well as from the SRDB references cited below: 

(1) It is questionable whether a “strict” contamination control program existed at the Burial 

Grounds, given evidence suggesting a haphazard approach to limiting contamination 

when dumping transuranic (TRU) waste drums, inadequate HP counting instrumentation, 

and apparent lax surveying of personnel and vehicles leaving the burial site (ANC 1972; 

SRDB Ref IDs: #141861, #125523).

(2) The apparent lack of a suitable smear counting capability at the Burial Grounds would 

have removed or severely impaired a key “workplace indicator” for indicating a potential 

intake and the need for a “special” bioassay. It is also not clear whether the HP program 

was sufficiently robust and independent at the Burial Grounds to support the necessary 

onsite surveillance to detect potential intakes. Pending the completion of an electronic 

database for INL bioassays, it is not possible at this time to ascertain the adequacy and 

completeness of that database and to what extent “special” bioassays figure in it (ANC 

1972; Burial Grounds 1962).

(3) Radioactive waste was not specifically identified for most drums, boxes, and other 

containers received at the Burial Grounds for disposal. A substantial amount of offsite 

waste was received from commercial, university, Energy Research and Development 

Administration (ERDA) and military sources when the NRTS was a national radioactive 

waste site; some of this waste was identified as to radionuclide content and activity 

levels, but much of it was not. It is not clear whether a suitable source term can be 

derived for what workers may have been exposed to from the standpoint of internal 

intakes and uptakes, and whether such exposures can be bounded by proposed NIOSH 

methods (INEL 1977; RWMC 1985; SRDB Ref IDs: #138957, #138704, #141859).

(4) The ER relies on a programmatic basis, i.e., strength of the HP program, to support its 

contention that no plutonium intake would have gone unmonitored. However, a number 

of investigations, program appraisals, and internal communications related to the Burial 

Grounds over the years have found fundamental shortcomings in that very program, 

particularly as it pertains to the detection, monitoring, and control of contamination. This 

record shows the AEC’s concern over the conflicted role of health physicists at the Burial 

Grounds, who were essentially responsible for much of its operation, as well as radiation 

protection. It also finds lack of management support for the Burial Grounds, generally, 

and to funding needed for contamination detection equipment, specifically. While some 

of these findings come from documents just after the 1970 cutoff in the current ER 

version, they nonetheless represent significant program deficiencies that transcend time 

periods at the Burial Grounds and raise serious doubt about the so-called “defense-indepth” approach taken by RWMC in the 1950s–1970s. In summary, given this checkered 

radiological program history, a programmatic basis alone is not sufficient to claim 

RWMC historic practices would have precluded any unmonitored plutonium uptake in 

the early years up to 1970 (ANC 1972; Burial Grounds 1962; SRDB Ref IDs: #141861, 

#138957, #141876).

The NIOSH ER concludes that worker exposures at the Burial Grounds can be dose 

reconstructed for 1952–1970 on the basis of stringent contamination controls, a radiation control 
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program exemplifying a “defense-in-depth” approach, and available internal dose data for known 

radioactive waste source terms that lend themselves to standard DR methods [e.g., OTIB-0054 

(ORAUT 2015) and OTIB-0060 (ORAUT 2007)]. SC&A believes all of these basic tenets are 

suspect, given a preliminary review of available SRDB documentation. 

Instead, SC&A believes available evidence indicates that the Burial Grounds was considered a 

low priority by INL management; was so underfunded that needed HP instrumentation was 

lacking; apparently lacked a management culture that supported disciplined operations and a 

formality of radiological controls to minimize unnecessary contamination; and dealt with high 

exposure potential MFPs and TRUs that were unidentified as to specific isotopic content, activity 

levels, and container content. From worker interviews, radiological incidents, and photographs 

of dumping operations, it is clear that an exposure potential existed for waste handlers. It is also 

clear that any resuspended contamination may not have been detected and necessary bioassay 

follow-up would not have occurred, given the management deficiencies at the site and the 

apparent lack of smear counting equipment available for Burial Grounds activities. 

Recommended Areas for Further Inquiry: 

 

(1) Investigate whether additional evidence (beyond worker interviews, incidents, and 

photographs) exists to support potential exposure of Burial Grounds waste handlers to 

radionuclide intakes. Were HP program practices, instrumentation, and dosimetry 

adequate to detect and identify intakes and uptakes? 

(2) Investigate how contamination control was administered at the Burial Grounds during the 

period 1952–1970. Were surface smears taken of personnel, equipment, and containers 

to identify loose contamination, and if so, how was that program affected by the apparent 

lack of smear counting equipment at the Burial Grounds (as identified in 1972)? Was 

alpha contamination given attention? 

(3) Investigate whether routine and special air sampling data are available for the Burial 

Grounds that can characterize the airborne particulate concentrations of MFPs, TRUs, 

and other resuspended contamination during disposal operations. Was there any air 

sampling performed after barrels or other containers broke open during disposal? 

(4) Investigate the effectiveness of the management of waste disposal operations at the Burial 

Grounds during this period, particularly the radiological controls program (i.e., did it 

constitute a “defense-in-depth” approach)? Was there any evidence of operational 

priorities taking precedence over those of radiation protection (HP)?

(5) Investigate the bioassay monitoring history at the Burial Grounds. When were routine invivo and in-vitro bioassays conducted for Burial Grounds workers and what “special” 

bioassays were conducted and for what reason? Is there a complete and adequate 

bioassay database to support dose reconstruction? Is it feasible to identify all INL 

workers who handled radioactive waste at the Burial Grounds (RWMC); what proportion 

of these workers were outside contractors or subcontractors? 

(6) Investigate whether information exists that enable source term characterization of all 

radioactive waste disposed in the Burial Grounds during 1952–1970. What radionuclides 

or mixed radioactive waste would not be covered under OTIB-0054 (ORAUT 2015) or 
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OTIB-0060 (ORAUT 2007)? How will unknown radioactive wastes be addressed in 

terms of dose assessment? Would MFPs, as a source term, dominate any other potential 

internal radiation exposure at the Burial Grounds? 

2.2.4 Central Facilities Area (CFA) 

SC&A conducted a review of the SEC Petition Evaluation SEC-00219 concerning radiation 

exposures and monitoring data available for the CFA at INL. NIOSH proposes bounding the 

internal doses from fission products by relying on Sr-90 and/or Cs-137 intakes in accordance 

with ORAUT-OTIB-0054 (ORAUT 2015) and ORAUT-OTIB-0060 (ORAUT 2007). Potential 

intakes of other radionuclides (i.e., uranium) in the presence of MFPs would be estimated on a 

case-by-case basis using the approach described in ORAUT-TKBS-0007-5 (ORAUT 2010). 

NIOSH believes this approach is sufficiently accurate to estimate the internal doses for CFA 

workers during the period from January 1, 1953, through December 31, 1970. 

According to the SEC, routine in-vitro monitoring was initiated in 1953, consisting of singlevoid urine samples. Gross beta analyses began in 1953, followed with gross gamma analyses 

beginning in 1957 (ORAUT 2010). According to Table 7-14 of the ER, in-vivo (WBC) 

monitoring did not begin until 1961. Routine in-vivo and in-vitro analyses were not performed 

for alpha-emitting radionuclides. 

SC&A has identified four CFA facilities that may not be suitable for the internal dose bounding 

approach proposed by NIOSH. They are listed in Table 4. 

Table 4. CFA Facilities of Concern 

Facility 

Number Facility Name Activity Description

CF-640 Machine Shop

Radioactive material that the plant shops could not handle was worked on in the 

more fully equipped CFA Machine Shop. Usually this material was of a low 

radiation and contamination level.

CF-665 Maintenance Shop

The Maintenance Shop personnel would sometimes work on vehicles and 

equipment that were used to haul radioactive material. Such vehicles were 

surveyed prior to shop maintenance work and, when necessary, sent to CPP for 

decontamination.

CF-669

Central Facilities 

Laundry 

(constructed in 

1950)

The laundry washed coveralls and other protective clothing items that were used 

in radiological work. The laundry drain went to a septic tank and drain field 

with other sanitary waste. The laundry facility and drain field(s) are sources of 

low-level radioactive contamination, which covers the spectrum inherent to 

work in radiological contamination areas. The old laundry facility (CF-699) was 

used from 1950 and was demolished in 1994. The new laundry facility 

(CF-617) was used from 1981 to 2001 and demolished in 2002.

CF-674 Sewage Treatment 

Plant

In the late 1950s and into the 1960s, small amounts of radioactivity were 

processed through a Sewage Treatment Plant, CF-674, to a drying pond. Most 

of the radioactivity was from the hot laundry, although small amounts could 

enter from CF-656 and CF-690.
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Each of the facilities listed in Table 4 would handle/process radioactive materials from the entire 

INL site, consisting of MFPs, mixed activation products (MAPs), actinides, or a mixture of any 

or all. For this reason alone, it would be difficult to use Sr-90 and/or Cs-137 ratios to determine 

MFP and MAP intakes, and not practical for actinide intakes, due to the uncertainty of the source 

term.

SC&A conducted a limited search of the SRDB for operations and survey information regarding 

the buildings identified in Table 1. Very little information was found concerning the operation 

and conditions in Buildings CF-640, CF-665, and CF-674. Several interoffice memos, reports, 

and radiological surveys were found for the Hot Laundry, CF-669. The concerns listed below, 

while specific to the CF-669 Hot Laundry, also apply to each of the other facilities in Table 4. 

Bioassay Monitoring 

There appears to be a lack of consistent and regular bioassay monitoring for MFP/MAP and no 

monitoring for uranium or other actinides. The routine urine bioassay sampling frequency was at 

most annual, and it is not clear if each employee was sampled annually or if a random group was 

sampled and considered to be representative of the workforce. The laundry was contaminated 

with both beta/gamma and alpha contamination, as indicated in the 1953–1956 contamination 

surveys (SRDB Ref ID: #139224). 

Air Monitoring/Ventilation 

The quality and adequacy of the air monitoring and ventilation in CF-699 is also questionable. 

According to an interoffice memo, “Laundry Improvements,” from November 19, 1971 (SRDB 

Ref ID: #143310): 

The contaminated air that is being released to the atmosphere from the five 

clothes dryers is of immediate concern. These dryers have two sets of lint 

removers, both of which are grossly inadequate for preventing contaminated lint 

or particles from escaping. Using an RM-14 Eberline instrument, we surveyed 

the first stage lint screens and the second stage lint screens and found 

contamination to 50,000 and 30,000 c/m, respectively. Using the same instrument 

we found the roof contaminated to a maximum of 25,000 c/m with an average of 

about 2000 c/m. The roof contamination appears to be on escaped lint. 

The same document also states: 

The constant air monitor (CAM) is worn out. 

There is no designed ventilation control system. Natural ventilation is wholly 

inadequate for this type of facility.

The areas handling "hot" laundry should be kept at negative pressure relative to 

the "clean" area. 

Hood ventilation should be provided for sorting incoming laundry.
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Contamination Control/Monitoring 

Concerns also exist regarding the quality and adequacy of contamination control and monitoring 

methods. According to a “CFA Health Physics Monthly Report for July 1961” (SRDB Ref ID: 

#139258), bagged laundry arrives without the proper contamination identification: 

Irregularities still occur at the radioactive laundry. Personnel become 

apprehensive when clothing comes in without tags or without Form ID-110 

indicating the contamination levels. 

There is also a concern with the contamination detection equipment. It was noted in the 1971 

memo (SRDB Ref ID: #143310): “Table monitors and clothing monitors are obsolete and 

inadequate. Proximity to "hot" laundry causes excessive background.” 

On a more general note regarding the state of the HP equipment, an interoffice memo dated April 

4, 1972 (ANC 1972) states in part: “The instruments now being used at the Central Facilities 

Health Physics office are outdated because of the reasons enumerated above, and because of 

their unreliability.” 

According to the April 4 memo, the HP smear counting equipment in use at the time (1972) 

consisted of a beta-gamma smear counter acquired in 1957, a beta-gamma smear counter 

acquired in 1958, and an alpha smear counter from 1952. The quality and reliability of the 

equipment in use impacts the quality and reliability of the contamination smear data from 

buildings, vehicles, shipments, and personnel monitoring in the CFA. 

SC&A believes that because of the concerns listed above and that CFA facilities handled 

radioactive materials from the entire INL site, it would be difficult to use Sr-90 and/or Cs-137 

ratios to determine MFP and MAP intakes, and not practical for actinide intakes due to the 

uncertainty of the source term. 

SC&A recommends that an evaluation should be conducted of the CFA radiological survey and 

air sampling results both during operations and just prior to decontamination and 

decommissioning (D&D) to determine actinide to Sr-90 and actinide to Cs-137 ratios. These 

ratios should then be compared to the values in Tables 5-22 and 5-23 of ORAUT 2010. 

2.2.5 Chemical Processing Plant (CPP) Pre-1963 

As noted in Section 1, the current SEC class definition as proposed by NIOSH includes all 

externally monitored workers at the CPP from January 1, 1963, to December 31, 1974. The 

period subsequent to the proposed SEC is still being evaluated by NIOSH, with a focus on the 

practical implementation of improved radiological safety policies beginning in 1975. For the 

period prior to the proposed SEC (pre-1963), NIOSH believes that sufficient information and 

monitoring records exist to reconstruct both internal and external doses to all relevant workers at 

the CPP. This section presents SC&A’s preliminary concerns and progress on evaluating 

potential SEC issues at the CPP prior to 1963. 
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Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Table 5 (reproduced from Table 7-5 of NIOSH 2015) shows NIOSH’s assessment of the 

feasibility for internal DR in different sources and time periods for CPP. Prior to 1963, NIOSH 

has determined internal doses can be reconstructed for MFPs, plutonium, neptunium, uranium, 

thorium, and any other relevant TRUs. The table also indicates that no coworker models are 

necessary prior to 1962. 

Table 5. Feasibility Summary for CPP (1953–1974)

Exposure

Source

Years

53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74

MFP F F F F F F F F F F F F F F C C C C C C C C

Pu/Np F F F F F F F F F F I I I I I I I I I I I I

Uranium F F F F F F F F F F I I I I I I I I I I I I

Th/Transuranics F F F F F F F F F F I I I I I I I I I I I I

Other F F F F F F F F F F F F F F C C C C C C C C

Photon F F F F F F F F F F F F F F F F F F F F F F

Neutron F F F F F F F F F F F F F F F F F F F F F F

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible, but a mixed-fission-product coworker model is needed. 

I = Dose reconstructions are infeasible.

Source: NIOSH 2015, Table 7-5 

Page 189 of NIOSH 2015 summarizes the rationale for the current SEC class definition as 

follows: 

Increased potential for intake due to poor contamination control and inadequate 

personnel monitoring for exposures to transuranics separated from mixed fission 

products makes it unlikely that exposures to alpha emitters can adequately be 

reconstructed from January 1963 through December 1974. 

Therefore, SC&A’s preliminary investigation into periods at CPP prior to 1963 focuses on two 

main facets: 

1. Characterization of contamination events and contamination control prior to 1963 and the 

associated implications concerning chronic exposure potential to CPP workers. 

2. Assessment of the available internal dosimetry program and its ability to capture and 

bound the potential internal exposure to CPP workers. This is particularly important in 

the assessment of internal exposure potential to alpha-emitting TRU isotopes. 

With regard to Item 1, SC&A began by inspecting relevant HP reports, which often contain 

summary characterizations of contamination identified within the CPP. In particular, SC&A has 

tabulated a preliminary data compilation concerning the number of contaminated surfaces 

identified by month at the CPP for periods prior to the proposed SEC class definition. This is 

shown in Figure 1. As seen in the figure, there are periods prior to 1963 in which there were a 

significant number of identified contamination events. In some cases, the number of 

contamination events appears to be higher than the first couple years of the SEC period. SC&A 

has not yet established information as to the magnitude and type of contamination events prior to 

1963. 
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0

50

100

150

200

250

Jul-56

Nov-56

Mar-57

Jul-57

Nov-57

Mar-58

Jul-58

Nov-58

Mar-59

Jul-59

Nov-59

Mar-60

Jul-60

Nov-60

Mar-61

Jul-61

Nov-61

Mar-62

Jul-62

Nov-62

Mar-63

Jul-63

Nov-63

Mar-64

Jul-64

Nov-64

Sample/Operating Corridors

Access Corridor

Figure 1. Number of Contaminated Surfaces in CPP-601 Sample/Operating Corridors 

and Access Corridor 

In addition to the standard periodic HP reports, SC&A has begun investigating specific incident 

reports in which contamination was discovered. In particular, information is being reviewed and 

compiled that can be used to characterize the magnitude of contamination, nature of event (offnormal or “unusual” events versus more common events), decontamination procedures 

undertaken, special and/or routine sampling, and steps taken to mitigate the occurrence of future 

events. 

With regard to Item 2, SC&A has begun investigating the files of claimants who have direct 

evidence of work at CPP prior to 1963 in order to characterize information concerning: 

 Internal personnel monitoring frequency and type (in vitro, in vivo, nasal swipes, etc.) 

 Evidence of contamination events or incident descriptions [includes both relevant 

Computer-Assisted Telephone Interview (CATI) information, as well as Department of 

Energy (DOE)-supplied monitoring records] 

 Any other relevant information pertinent towards characterizing the internal exposure 

potential and monitoring program at CPP prior to 1963 

In addition to the review of claimant-specific records, hardcopy bioassay records exist on the 

SRDB that will help characterize the internal monitoring program as it relates to exposures at 

CPP. Currently, these bioassay data only exist in a complete form as hardcopy records; however, 

a partial electronic compilation of such records is available. Such a compilation is somewhat 

useful in making preliminary judgments as to the direction of future investigative lines of 

inquiry, with the stipulation that such records are not currently complete, but will likely become 

complete at a future date. 
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Recommendations: The combination of information gathered for the items described above will 

allow SC&A to evaluate the extent to which internal exposure potential existed for the relevant 

worker population at CPP, as well as the sufficiency of the internal monitoring program at CPP 

to capture all relevant exposures. SC&A recommends that the following review actions continue 

to fully assess the ability to reconstruct internal doses at CPP prior to 1963: 

 Continue research and data collection to characterize the number, location, and 

magnitude of contamination events/incidents at CPP prior to 1963. 

 Evaluate actions taken both prior to, during, and after contamination events/incidents to 

assess the radiological control program in place at CPP during the earlier period. 

 Assess the extent to which chronic lower-level contamination may have existed at CPP 

(beyond noted incidents and contamination events) that may have presented a long-term 

chronic internal exposure potential to CPP workers. 

 Review individual claimant dosimetry and other records available for CPP workers prior 

to 1963 to characterize the extent of the routine and incident-based dosimetry program as 

it relates to internal doses from all relevant contaminants present at CPP. 

 Evaluate current hardcopy bioassay data, to the extent feasible, for completeness and 

adequacy bounding internal doses at CPP. Additionally, evaluate any completed 

electronic database records as they become available. 

 Research changes in production processes and amount of materials processed that could 

provide an indication of the type of contamination that was occurring. 

2.2.6 Test Area North (TAN) 

Because of the complexity of the operations at TAN, it is informative to establish a work 

breakdown structure based on the different operations and sub-operations that occurred at that 

site. The following is an attempt to create such a structure, which can be used to help organize 

an approach to evaluate data completeness, data quality, and data adequacy for the broad range 

of operations, research activities, and campaigns that took place at TAN. 

 Aircraft Nuclear Propulsion Program (ANP) (1952–1961) 

 Initial Engine Test (IET) – This may be a subset of ANP and also a stand-alone program 

o Heat Transfer Reactor Experiments 

 Technical Support Facility (TSF) 

o TAN 607 Hot Shop

o LOFT (TAN 650) 

o Storage Pool 

o Storage Pads (TAN 790 and 791) 

o Radwaste Liquid Disposal System 

o Storage Building 

o Radiography Facility (TAN 607) 
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 Water Reactor Research Test Facility (WRRTF) 

o Low Power Test Facility (LPTF) 

o Shield Test Pool Facility (STPF) 

 Specific Manufacturing Capability (SMC) – DU machining for armor 

Table 5-4 in the SEC ER provides a breakdown and description of each of these facilities and 

associated radiological issues. Our initial overarching observations regarding the external 

dosimetry programs at these facilities are favorable, with the possible exception of neutron 

dosimetry. Possible concerns regarding the internal dosimetry program as described above also 

apply to the TAN facilities. 

Given (1) the complexity of the activities at TAN, (2) the attention that has been given to the 

internal dosimetry programs at INL in general, and (3) the apparent quality of the external 

dosimetry data collected at TAN (with the possible exception of neutron dosimetry), this section 

of the report focuses on an initial review of the external dosimetry records for TAN, as part of a 

larger effort to review the SEC ER. 

The following analysis was limited to records available on the SRDB. This investigation focuses 

on determining whether the external dosimetry data for workers at TAN are of sufficient quality 

and completeness. The level of detail given regarding the various types of dosimeters used, as 

well as calibration practices in ORAUT-TKBS-0007-6 (ORAUT 2011), indicates that the 

external dosimetry data appear to be high quality. Overall, it was seen that the external 

dosimetry records for TAN as a whole (including all sub-areas) appear to be fairly complete, 

spanning from mid-1955 through part of 1970 with a gap from June to December of 1961. 

However, when attempting to separate the data into the various subdivisions of TAN, the data do 

not appear to be very complete, or the labeling of the records does not always provide 

information that can place a worker at a given facility within TAN at the time of the readout. 

The available dosimetry records often indicate a sub-area of TAN, but very limited dosimetry 

records were found for some areas (e.g., STPF). Also, the dosimetry reports do not indicate 

specific buildings within TAN for workers, only a general area. 

In order to begin investigations into external data completeness at the various TAN facilities, 

SC&A performed numerous searches within the SRDB in an effort to capture as many 

documents as possible that contained dosimetry information. Examples of search terms used 

include dosimetry, dosimeter, external, personnel, badge, and film. For each of these documents, 

the date of each dosimeter cycle, the areas or area codes mentioned, and the number of neutron 

badges were recorded. In some instances, the document would primarily contain dosimeters for 

one area (e.g., ANP), but some workers would have multiple badges that may include other areas 

or area codes (e.g., STPF). Any records of this type were noted and recorded in this analysis. 

Characterization and Commentary on Neutron Dosimetry Data 

The manner in which neutron badges were reported in these dosimetry documents differs 

depending on the year and format of the report. In the older records, neutron badge readouts 

were handwritten in the “remarks” section of the report. Within these types of records, this 

notation would sometimes have the word “neutron” or just an “n” before the dosimeter reading 
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of “xx mrem.” Therefore, when the remarks section would only say “xx mrem,” it was also 

taken to mean a neutron badge reading. Other neutron badge data were formatted with separate 

columns for neutron results and were thus much easier to read. These types of records were 

primarily seen for the early to mid-1960s. The final way in which neutron badges were 

presented had summary pages for each badging period, which helped to infer which columns in 

these reports corresponded to beta, gamma, and neutron readings. Even though there are 00s 

reported in the neutron column in some of the records, these cannot be taken to mean a zero 

neutron reading. It was found that if a worker had a neutron badge reading in a given period, 

then a second set of numbers would be present in the neutron column in the rows associated with 

that worker. 

Commentary on Completeness of External Dosimetry Data at TAN 

Table 6 gives information about each document containing detailed dosimetry records that were 

analyzed for this initial completeness analysis. The number of badges per document was taken 

to be an estimate of the average number of badges per page, multiplied by the number of pages 

in the document. These are likely to be over-estimates, as it was found that some pages within 

the documents were summaries, had zero entries, visitor badges, or contained badges that were 

not read out, often accompanied by an irregularity code of 14 (Not in area). The number of 

neutron badges in each document only takes into account those reported for badges associated 

with an area of TAN, and were not from visitor badges. 

Table 6. Documents Found for TAN in SRDB with Detailed Dosimetry Records

SRDB Reference ID # Area(s) Mentioned # Pages # Estimated 

Badges

# Neutron 

Badges

116020 IET/STEP 535 8025 226

116021 TAN GE (ANP) 376 3760 231

116022 TAN GE (ANP) 387 3096 185

116023 TAN GE (ANP), LPTF4, STPF1125 2625 0

116030 TAN GE (ANP), STPF1264 4488 3625

116031 TAN GE (ANP) 561 5610 396

116032 TAN GE (ANP) 373 3357 201

116033 TAN GE (ANP) 612 5508 312

116034 TAN GE (ANP) 337 3033 125

116035 TAN PPCo, TSF 525 9450 3

116036 TAN GE (ANP) 464 4176 294

116037 TAN GE (ANP) 393 3930 250

116038 TAN GE (ANP) 483 4830 106

116039 TAN GE (ANP) 476 3332 117

1416366 TAN GE (ANP), LPTF, IET/STEP 24 264 0

119002 TSF 671 6710 2

119005 GE (ANP) 213 5325 309

 

4

 These areas are present in some workers’ multiple badges within a report for a different area, not as a 

separate report for the area. 

5

 All neutron badges are associated with workers’ badges for STPF. 

6

 Document is titled as an ‘excerpt;’ some pages are contained within other documents listed. 
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Table 6. Documents Found for TAN in SRDB with Detailed Dosimetry Records 

SRDB Reference ID # Area(s) Mentioned # Pages # Estimated 

Badges

# Neutron 

Badges

119006 GE (ANP) 182 5460 173

119017 GE (ANP) 221 5525 50

119018 GE (ANP) 309 4635 477

119019 GE (ANP) 141 2115 280

119020 GE (ANP) 233 6990 319

119030 GE (ANP) 286 7150 138

119031 TSF 168 2688 1

119032 TSF 493 12325 0

119033 TAN PPCo 474 11850 2

119034 GE (ANP) 236 5900 424

119035 GE (ANP) 117 2925 236

119036 GE (ANP) 388 9700 645

119037 GE (ANP) 166 4150 239

119038 GE (ANP) 341 8525 101

119039 GE (ANP) 241 4820 184

119040 GE (ANP) 474 11850 55

119041 GE (ANP) 293 7325 409

119042 GE (ANP) 405 10125 175

119043 TSF 73 1460 0

119044 TSF 117 1872 0

Overarching Observations Regarding the Completeness of the External Dosimetry Data at TAN 

In order to assess the completeness of data for TAN as a whole, the dosimeter dates for all of the 

documents listed in Table 6 were combined, regardless of the TAN sub-area the dosimeters 

might have been associated with. Figure 2 is a graphical representation of the dates for which 

dosimeters exist for an area of TAN that are contained within the analyzed SRDB documents. 

The y-axis also represents the year of the dosimeter, such that the data could be broken up by 

year to allow for easier identification of temporal gaps in the available records. The points that 

make up the section of Figure 2 for 1956 are shown in more detail by themselves in Figure 3. 

What looks like an irregularity in the dosimeter dates for 1956 could be attributed to non-routine 

readouts; for example, if a badge was designated “late-pull” or “HP request,” then a date usually 

different than those of the typical dosimeter cycles was recorded. 
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1955

1956

1957

1958

1959

1960

1961

1962

1963

1964

1965

1966

1967

1968

1969

1970

Figure 2. Dates of Dosimeters Available in SRDB Documents for All of TAN 

0

1

Figure 3. Dates of Dosimeters for 1956 taken from Figure 2 

Overall, most years appear to have dosimeter cycles that reasonably cover that entire year, with 

the possible exceptions of 1955, 1961, and 1970. The earliest dosimeter cycle within the 

documents was dated April 10, 1955. The available dosimeter records in 1961 do not cover 

June 18 through December 27. Dosimeters for the month of August in 1970 were also not found. 

In SRDB Ref ID: #119002, the dosimeters appear to be reported on a monthly basis, but skip 

from July 31 to September 1, 1970. Investigation into whether the records for these gaps exist 
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may be necessary. It is also possible that the records exist in the SRDB, but were not captured 

during the searches.

Completeness of External Dosimetry data for Sub-Areas of TAN 

Due to the fact that most of the dosimeter records either had an area written into its header or had 

area codes associated with the dosimeters, an effort was made to see if the data are complete 

enough to break the dosimetry records into sub-areas of TAN. Figure 4 illustrates the dates of 

dosimeters available for various sub-areas of TAN within the analyzed SRDB documents. 

0

1

2

3

4

5

6

1/1/19551/1/19561/1/19571/1/19581/1/19591/1/19601/1/19611/1/19621/1/19631/1/19641/1/19651/1/19661/1/19671/1/19681/1/19691/1/1970

1 = ANP

2 = IET/STEP

3 = STPF

4 = LPTF

5 = TAN Only

6 = TSF

Figure 4. Dates of Dosimeters for Various Sub-Areas of TAN 

Other Evidence of Badging 

While searching the SRDB for dosimetry records, many documents were found that contained 

monthly or annual summary reports, as well as records containing only visitor badges. 

Considering TAN as a whole, Figure 5 shows the number of each type of document found in the 

SRDB that covers at least part of a given year. If a document covered multiple years, it was 

counted for each year. It can be seen that summary and/or visitor badge reports are available 

outside of the time period for which detailed records are available (1955–1970). Table 7

presents a list of the summary and visitor reports found. 
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1 

3 

5 6 6 5 

2 

6 

3 

6 

3 

7 

3 3 3 

1 

1 

1 

1 

1 1 

1 

1 

1 

1 

1 

1 

1 

1 1 1 

2 

1 2 1 1 1 

4 

1 

2 

2 

4 

3 

5 

2 

1 1 1 

1 

1 

1 

1 1 

0

2

4

6

8

10

12

1955 1956 1957 1958 1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972 1973 1974 1975

Detailed Reports Summary Reports Visitor Badge Reports

Figure 5. Number and Type of Documents Found in the SRDB According to the 

Year(s) They Cover 

Table 7. Visitor or Summary Documents Found for TAN in SRDB 

SRDB Reference ID# Summary Report or Visitor Badge Report

116024 Visitor

116025 Visitor

116026 Visitor

116027 Visitor

116028 Visitor

116029 Visitor

116040 Visitor

116041 Visitor

118967 Visitor

118970 Visitor

118982 Visitor

118983 Visitor

118989 Visitor

118990 Visitor

116042 Visitor

118965 Visitor

118968 Visitor

118985 Visitor

118986 Visitor

118987 Visitor

118988 Visitor

138150 Visitor

138152 Visitor

138153 Visitor

125917 Summary

125913 Summary

125915 Summary

125920 Summary

125924 Summary

125926 Summary

125927 Summary

125928 Summary

125930 Summary

125934 Summary
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Summary Reports 

Several of the summary reports give the number of badges, pencil dosimeters, neutron badges 

(A-film), as well as other types of dosimeters read for a given month for various areas of INL. In 

one summary report (SRDB Ref ID: #125917), the first mention of an area associated with TAN 

was in April of 1955 (as ANP). This coincides with the earliest dosimeters associated with ANP, 

which were also in April of 1955. Use of these summary reports may be useful regarding the 

assessment of the completeness of the available dosimetry records. For example, in SRDB Ref 

ID: #125917, the number of neutron badges recorded for October of 1957 for ANP is reported as 

186 (p. 236). However, the number of badges counted within the analyzed detailed reports for 

ANP during this month is 175. Further investigation and comparison between these records 

might be necessary to know whether more dosimetry records exist, and what types of workers 

they are attributed to. Some of the summary reports give the number of visitors to each area, but 

it is unclear whether any visitor badges are included in the reported badge readings for a given 

month and year. In addition, these summary-style records sometimes also include categories 

representing dosimeters used for construction workers. 

NIOSH does have an electronic database containing external dosimetry records as well. 

However, the records are given as an annual summary per person, and have no information 

regarding the area(s) where the worker may have received the dose. From an initial glance at the 

external dosimetry database, it does appear that the data span from the 1950s through at least the 

1990s, and contain hundreds of thousands of rows of data, yet the absence of area information 

limits the possible uses of these data. 

Recommendations for Further Investigations: 

1. Figures 2 and 3 identify a rather complete external dosimetry set of records for TAN, 

with only a limited number of gaps. Additional SRDB searches may help fill these gaps. 

2. The external dosimetry records often do not provide information regarding the sub-areas 

where a given worker experienced exposure during a given change-out. In light of the 

highly varied activities that took place at the different sub-areas, the complete TAN 

dataset cannot be used to build a coworker model for unmonitored workers at a given 

facility within TAN. It is therefore recommended that the data available for each subfacility be compiled and plotted for use in building a coworker model for each facility, if 

it is determined that coworker models are needed due to the incompleteness of the 

external dosimetry database. 
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3.0 REFERENCES 

ANC 1972. Interoffice Correspondence from C.E. Nichols to R.B. O’Brien, Subject: Central 

Facilities and Burial Ground Smear Counters. Aerojet Nuclear Company. April 3, 1972. SRDB 

Ref ID #143623. 

Burial Grounds 1962. NRTS Burial Ground Standard Practices Manual; 5.61.2; author 

unknown; SRDB Ref ID #140874, PDF p. 3.

Croff 1980. Croff, A.G., ORIGEN2 – A Revised and Updated Version of the Oak Ridge Isotope 

Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory, July 1980. 

SRDB Ref. ID: #23569. 

INC 1971. Health Physics Appraisal, Idaho Nuclear Corporation, May 11, 1971. SRDB Ref ID: 

#116116. 

INEL 1977. Waste Management Program – History of Buried Transuranic Waste at INEL, 

WMP 77-3, Idaho National Engineering Laboratory, Energy Research and Development 

Administration, March 1977. SRDB Ref ID: #082970. 

NIOSH 2015. SEC Petition Evaluation Report: Petition SEC-00219, National Institute for 

Occupational Safety and Health, Cincinnati, Ohio. March 12, 2015. 

ORAUT 2007. Internal Dose Reconstruction, ORAUT-OTIB-0060, Rev. 00, Oak Ridge 

Associated Universities Team, Cincinnati, Ohio. February 6, 2007. 

ORAUT 2010. Technical Basis Document for the Idaho Engineering and Environmental 

Laboratory (INEEL) – Occupational Internal Dose, ORAUT-TKBS-0007-5, Rev. 03, Oak Ridge 

Associated Universities Team, Cincinnati, Ohio. March 2, 2010. 

ORAUT 2011. Technical Basis Document for the Idaho National Laboratory and Argonne 

National Laboratory-West – Occupational External Dosimetry, ORAUT-TKBS-0007-6, Rev. 03, 

Oak Ridge Associated Universities Team, Cincinnati, Ohio. SRDB Ref. ID: #94104.

ORAUT 2014. Supporting Calculations for ORAUT-OTIB-0054 and ORAUT-RPRT-0047, 

ORAUT-RPRT-0067, Rev. 00, Oak Ridge Associated Universities Team, Cincinnati, Ohio. 

August 26, 2014. 

ORAUT 2015. Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, Rev. 03, Oak Ridge Associated 

Universities Team, Cincinnati, Ohio. February 6, 2015. 

ORNL 2015. http://scale.ornl.gov/ (accessed 2/6/15) 

RWMC 1985 A History of the Radioactive Waste Management Complex at the Idaho National 

Engineering Laboratory; WM-F1-81-003, Rev. 3; author unknown; Waste Program Division, 
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Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Radioactive Waste Management Complex (RWMC) Operations Branch; July 1985; SRDB Ref 

IDs: #83913 and #124478.

SC&A 2015. SC&A Evaluation of Idaho Chemical Processing Plant Class Definition Requiring 

Evidence of External Dosimetry (1963–1974), SC&A, Inc., Vienna, Virginia, and Saliant, Inc., 

Jefferson, Maryland. June 29, 2015. 

SRDB Ref. ID: #119002 – This file contains 674 pages, most or all of which appear to be area 

exposure reports. 

SRDB Ref. ID: #125523 – Survey of NRTS Roads, Idaho Nuclear Corporation, December 30, 

1969.

SRDB Ref. ID: #125917 – This file contains 414 pages of Dosimetry Branch Activities 1952–

1959 (Sheets marked “Personnel Meters Behavior”) 

SRDB Ref. ID: #138704 – CFA Monthly Reports, 1963 (contamination surveys, burial grounds, 

HP).

SRDB Ref. ID: #138957 – Possible Internal Intake Report, August 23, 2006.

SRDB Ref. ID: #139224. CFA Contamination Survey Results 1953–1956.

SRDB Ref. ID: #139258. CFA Health Physics Monthly Reports for January 1959–December 

1961.

SRDB Ref. ID: #141608 – CFA HP – Selected Safe Work Permits for 1965.

SRDB Ref. IDs: #141859 and #141861 – Internal memorandum from G. Wehmann (Chief, HP 

Section) to W.P. Gammill (Chief, Site Survey Branch), Subject: Health Physics Practices at the 

NRTS Burial Ground. May 16, 1962. 

SRDB Ref. ID: #141876 – Letter from AEC to Aerojet Nuclear Company, Subject: Personnel 

Exposure at Burial Ground. September 20, 1973. 

SRDB Ref. ID: #143310. Laundry Improvements, Interoffice Correspondence from O’Brian to 

Lyon, November 19, 1971. 

SRDB Ref. ID: #143623. Central Facilities HP Instruments, Aeroject Nuclear Company 

Interoffice Correspondence from O’Brian to Lyon, Nichols to Lyon, April 1972. 

SWP 1962–1964. Safe Work Permit Burial Ground January 1962 – December 1964; author 

unknown; 1962–1964; SRDB Ref ID: #138194.